Infinitely dilute, unbroadened point cross sections are obtained from resolved resonance parameters using a modified version of the RESEND program (3) (NESC0465). The SIGMA1 (4) (IAEA0854) kernel-broadening method is used to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0 K (outside of the unresolved energy region). Effective temperature- dependent self-shielded pointwise cross sections are derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation intervals. The selection of energy mesh points, at which the effective cross sections are calculated, has been modified to include the energy points given in the ENDF/B file or, if the energy-independent formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral shape under user control. The integral over energy for each group is divided into a set of panels defined by the union of the grid points describing the total cross section, the reaction cross section of interest and the gross spectral shape. The integration scheme actually employed in MINX is adaptive Simpson's procedure for which the initial estimate is based on the unionized grid described above. The computation of elastic and discrete group- to-group matrices is based upon a semi-analytic scheme which treats the rapidly fluctuating cross-section behaviour analytically. Where this laboratory-system-based scheme becomes difficult to implement (e.g., light nuclei, inelastic thresholds), an alternative numerical integration in the center-of mass system is employed. Multigroup transfer matrices for processes in which the outgoing neutron energy and angular distribution is uncoupled are computed by direct numerical integration.

The principal restriction is the computing time available for a given desired accuracy, number of groups, and Legendre order. The paging technique and variable dimensioning make efficient use of available core storage; very large problems have been run with MINX (e.g. a complete 171-group P3 neutron library at ORNL and an extensive 240-group P4 library at LASL).

MINX generates and uses (in the resolved energy region) a linearly interpolable, infinitely dilute temperature-dependent pointwise cross-section library in ENDF/B-IV format. This feature permits efficient computation of group cross sections with accurate Doppler broadening of single-level and multi- level cross sections. The multigroup constants generated therefrom are thus known to be compatible with the pointwise cross sections retrieved by contuous-energy Monte Carlo codes. New, accurate algorithms for the computation of Legendre moments of group-to-group transfer matrices have been developed and implemented. These calculations are based on an expansion of the differential scattering cross section in the laboratory system and use a semi- analytic procedure which treats the rapidly fluctuating cross- section behaviour analytically. Where Legendre expansion in the lab system becomes difficult (e.g. for light nuclei or near inelastic thresholds) an alternative numerical integration in the centre-of- mass is employed. The procedures employed in MINX for constructing, interpolating and integrating cross sections are intended to provide and quantify user control of computational errors (assuming that the data base and weighting function are known explicitly). A paging technique which manipulates huge amounts of cross-section information one block at a time (block size variable), is used throughout MINX, in addition to variable dimensioning to reduce storage requirements and to use available storage efficiently. Finally, the code and the multigroup data sets derived therefrom are intended to satisfy nuclear design standards currently being implemented under auspicies of the American National Standards Institute.

Three utility codes are provided to manipulate the CCCC data files:
LINX: will combine two multi-isotope CCCC files (ISOTXS or BRKOXS
only) (6) (PSR-0129)
BINX: will convert CCCC data (ISOTXS, BRKOXS, or DLAYXS) from binary to BCD mode or vice-versa and selectively print the contents
of the files (6) (PSR-0129).
CINX: will exactly collapse fine group data (ISOTXS, BRKOXS, or
DLAYXS) to a subset coarse group structure, and will also
change the format of the data to 1DX/PERT-V form, if desired
(7) (PSR-0117).

These codes were written to be as IBM-compatible as possible. The changes required are identified on the listings with "C IBM comment" cards.

Core and external storage such as magnetic tape or disk devices depend on the characteristics of the problem. The 50-group library would require about 330k bytes of core storage on the IBM 360 series computer or 49000 words of LCM on the CDC-7600.

The program consists of 227 subroutines on about 15000 source cards. With a three-level overlay structure consisting of 11 segments, about 330k bytes are required on the IBM machines and 49000 decimal words on CDC-7600.